This invention relates to a method of treating spent fuel utilizable in a spent nuclear fuel retreatment process, scrap nuclear fuel wet reclamation process, etc.
Ordinarily, in spent nuclear fuel re-treatment and scrap nuclear fuel wet reclamation processes, organic solvent used in an extraction process is degraded by the effects of acidity and radiation. Consequently, the degraded products are removed from the organic solvent by a solution of sodium hydroxide or sodium carbonate, after which the solvent is reused.
Certain shortcomings, however, exist in such conventional methods. These are as follows:
(1) Reclamation of organic solvent in which there is advanced deterioration is impossible, and the solvent becomes a liquid radioactive waste that is difficult to treat.
(2) A solution containing sodium is mixed with radioactive liquid waste of the nitrate family, after which the resulting solution is reduced in volume and solidified in glass or asphalt. However, owing to the large amount of sodium contained, the reduction in volume has its limitations. This also accounts for complicated solidification treatments.
In view of the foregoing, there is a need to develop a process which minimizes the use of sodium as well as a solvent reclamation process.
Further, though evaporation cans are used to concentrate radioactive material in treatment of liquid radioactive wastes, these are disadvantageous because decontamination is inefficient and the cans are subject to considerable corrosion. It is desired, therefore, that a treatment process with a higher decontaminating efficiency and less corrosion be developed.